The testing of tensile strength and bursting strength of tubular products presents difficulties due to the number of types, sizes and uses for tubing. High temperature and pressure, usually with corrosion and shock loading, are problems in transportation, power generation, petrochemical systems. A particular concern is the cladding for fuel rods in nuclear power reactors including the many light water reactors used in Western countries.
The preferred material for nuclear fuel rod cladding in light water reactors is currently Zircaloy-4, manufactured by Sandvik Special Metals Corporation, Division of Sandvik Nuclear AB, Nykoping, SE. The basic design consideration in fuel rod designs are succinctly described in U.S. Pat. No. 4,004,972 to Mogard and in U.S. Pat. No. 4,822,557, also to Mogard. The discussions therein refer to conventional fuels at low burn-up (e.g., less than 60 GWd/MTU). Higher burn-up rates are desirable for lower net fuel costs. MOX (mixed uranium oxide, weapons grade plutonium oxide) is contemplated in the future for the disposal of surplus weapons grade plutonium, and its use entails both engineering and political questions.
Mechanical testing of fuel rod cladding using axial and circumferential load tests have been reported by R. S. Daum et at, “Mechanical Properties of Irradiated Zircaloy-4 for Dry Cask Storage Conditions and Accidents” Nuclear Safety Research Conference, Washington D.C. (2003), R. S. Daum et al., “Mechanical Property Testing of Irradiated Zircalog Cladding Under Reactor Transient Conditions,” Small Specimen Test Techniques: Fourth Volume, ASTM STP 1418, M. A. Solokov, J. D. Landes and G. E. Lucas eds, ASTM, West Conshohockey Pa. (2002), and Link et al., Nuclear Engineering and Design, 186, 379 (1998).
Methods using split rings to apply radial pressure include S. Uchikawa, “Ring Tensile Testing of Zircaloy Cladding Tubes at JAERT,” FSRM, Tokyo (2004) and ASTM D 2290-92 (1992). A test described as Expansion Due to Compression Test was reported by Dufourneaud et al., World Congress on Computational Mechanics, Vienna (2002) uses a polymer pellet axially compressed inside a tube by two pistons pressing at opposite sides.
The U.S. Department of Energy (DOE) Fissile Materials Disposition Program (FMDP) is pursuing reactor irradiation of mixed uranium-plutonium oxide (MOX) fuel for disposal of surplus weapons-usable plutonium. To pursue disposition of surplus weapons-usable plutonium via reactor irradiation, it must be demonstrated that the unique properties of the surplus weapons-derived or weapons-grade (WG) plutonium do not compromise the applicability of this MOX experience base.
One question to be addressed for weapons-derived MOX fuel is that of ductility loss of the cladding during irradiation. While irradiation induced loss of ductility has long been known and quantified for many cladding materials, the potential synergistic effects of irradiation and the unique constituents (i.e., gallium) of weapons-derived MOX fuel are not known. The Postirradiation Clad Ductility test Program formulated for DOE is conducted by the Oak Ridge National Laboratory (ORNL). The program focus is on development, validation, and application of technology for the determination of the residual ductility for mixed oxide (MOX) fuel cladding irradiated in the Advanced Test Reactor (ATR). The scope of the project includes development of techniques for machining and handling of small ring-type test specimens, development and validation of a specimen and test fixture for use in a hot cell environment, and testing of cladding specimens subjected to burnup levels from zero to more than 50 GWd/MT.
The test methods described above (e.g. ASTM D 2290–92) are applicable to clean rods and samples which have been scrupulously cleaned. They are bench tests and not safe or suitable for routine, reliable data collection in a hot cell environment where manipulation is limited.